The present invention relates to a method for measuring the coolant flow rate of a fuel element and to apparatus for practicing the method.
In water or liquid metal cooled reactors, each fuel element comprises a plurality of metal clad fuel rods which are arranged in a common wrapper tube in such a manner that a cooling channel is formed which has a cross section that is sufficient for removing the thermal energy. The operational safety of the reactor and the optimum performance of each fuel element is critically dependent on the mass flow rae with which the coolant flows through its cooling channel with respect to the power produced in the fuel element. Thus, it is necessary to continuously measure the coolant flow rate and the power of each fuel element, and to perform these measurements redundtantly so as to exclude erroneous operations.
Although flow meters for measuring the rate of flow of liquid metals are known, their installation is generally limited to locations in the main coolant lines. This limitation is due to their complicated design, tendency to malfunction, relatively large space requirement and limited service life.
It is also known to monitor the coolant flow rate of reactors indirectly by continuously measuring the coolant exit temperature at the fuel element outlet. However, the coolant exit temperature depends not only on the coolant flow rate but also on the fuel element power which is influenced by the operating conditions and the duration of operation.
It is further known from the publication Atomwirtschaft (Atomic Energy Economics), December 1973, pages 580-582 that the velocity of vapor bubbles developing in the coolant of boiling water reactors may be determined by correlation methods. This technique requires a plurality of neutron detectors which are arranged within the fuel element one on top of the other when seen in the direction of flow. The actual coolant velocity, however, is less than the velocity measured by this method due to slippage of the vapor bubbles, and the method can be used only with the two-phase streams found in boiling water reactors.
In another prior art measuring method of this type, which is discussed in the book, Progress in Nuclear Energy, Volume 1, pp. 553-563, Pergamon Press 1977, Vol. 1, two thermocouples spaced at a distance of about 10 cm are positioned in a closed flow channel in the direction of coolant flow.
It is an object of the present invention to provide a method and apparatus which make it possible, without additional incore instruments and without using flow meters, to continuously measure the coolant flow rate of every fuel element and to detect cooling malfunctions at the time they develop. This permits malfunctions to be immediately corrected thereby preventing seriou damage to the reactor.